Phase change in the nuclear reactor core is related to safety criteria such as Departure of Nucleate Boiling (DNB) during normal and transient conditions. So that, a lot of computer codes with verification and validation against experiment are used to investigation of thermal hydraulics behavior of vertical boiling flow in core channel with system and component scales. Until now, even many studies on boiling flow are implemented in CFD scale codes, but their utilization to specific nuclear reactor is not yet applied. Thus, the utilization of many codes including CFD scale (Ansys CFX) to investigate void fraction in hot channel of VVER- 1000/V392 reactor core is studied in this work. Due to VVER-1000/V392 nuclear reactor is a candidate for Ninh Thuan 1 nuclear power project, so that the understanding of VVER’s reactor technologies including research works of this thesis is important to develop competence of nuclear safety in Vietnam.